ORAU Team

NIOSH Dose Reconstruction Project

Technical Information Bulletin – Maximum Internal Dose 

Estimates for Certain DOE Complex Claims 

Document Number:

ORAUT-OTIB-0002

Effective Date:11/07/2003

Revision No.: 00

Controlled Copy No.: ________

Page 1 of 15

Subject Expert: Gene Rollins

Approval: Signature on File_______________ Date: 11/06/2003

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File____________ Date: 11/06/2003

Richard E. Toohey, Project Director

Approval: Signature on File_______________ Date: 11/07/2003

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 10/08/2003 00-A New document to establish Technical Information 

Bulletin for Maximum Internal Dose Estimates for 

Certain DOE Complex Claims. Initiated by Gene 

Rollins. 

11/07/2003 11/07/2003 00 First approved issue. Initiated by Gene Rollins.
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 2 of 15

MAXIMUM INTERNAL DOSE ESTIMATES FOR CERTAIN DOE COMPLEX CLAIMS

1.0

The purpose of this Technical Information Bulletin (TIB) is to provide a method to facilitate timely 

processing of claims under the Energy Employee Occupational Illness Compensation Program Act 

(EEOICPA 2000), which involve cancer to an organ with little or no reported internal dose from 

internally deposited radionuclides that might be associated with work at DOE complex sites. 

Purpose

2.0

Title 42, Part 82 of the Code of Federal Regulations (CFR) dictates the methods to be used for 

radiation dose reconstruction under the Energy Employees Occupational Illness Compensation 

Program Act of 2000. Section 82.10(k) summarizes the general philosophy to be adopted:

Background

“Research and analysis will be determined sufficient if one of the following three conditions is 

met:

(1) From acquired experience, it is evident the estimated cumulative dose is sufficient to 

qualify the claimant for compensation (i.e., the dose produces a probability of causation of 

50% or greater);

(2) Dose is determined using worst case assumptions related to radiation exposure and intake, 

to substitute for further research and analyses;

(3) Research and analysis indicated under steps described in paragraphs (f) - (j) of this section 

have been completed.

Worst-case assumptions will be employed under condition 2 to limit further research and 

analysis only for claims for which it is evident that further research and analysis will not 

produce a compensable level of radiation dose (a dose producing a probability of causation of 

50% or greater), because using worst-case assumptions it can be determined that the 

employee could not have incurred a compensable level of radiation dose.”

“Worst-case assumption” is defined in Section 82.5(r) as:

“A term used to describe a type of assumption used in certain instances for certain dose 

reconstruction conducted under this rule [42 CFR 82]. It assigns the highest reasonably 

possible value, based on reliable science, documented experience, and relevant data to a 

radiation dose of a covered employee.”

A number of EEOICPA claims have been submitted that can be expedited based on paragraph (2) 

above. Cases were reviewed to identify those with cancer of an organ that does not concentrate any 

of the primary radionuclides expected to be found at DOE sites and those with little or no apparent 

internal dose to that organ. 

In accordance with OCAS-IG-002 (Internal Dose Reconstruction Implementation Guideline, Rev. 0, 

August 2002), internal dose is assigned to employees who were monitored but had no detectable 

activity (“positive”) in their samples and to employees who were not included in a bioassay program, 
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 3 of 15

because there is some amount of intake and associated dose that is not detectable by an internal 

dosimetry program. To expedite the dose reconstructions, cases that met the criteria above and are 

thus considered unlikely to have significant internal radiation exposure, will first be evaluated with the 

method described in this TIB. If the outcome yields a probability of causation >50%, a dose 

reconstruction using more reasonable assumptions will be performed.

3.0

3.1 ASSUMPTIONS

Implausible Undiscovered Intakes

For claims where it is considered likely that the covered employee had no significant internal radiation 

exposure, a method to expedite claims has been developed in accordance with 42 CFR 82.10(k)(2). 

This method assumes the “largest reasonably possible value” of the source term comprised of 

radionuclides that are/were typically the more significant radionuclides (by either preponderance or by 

internal dose significance) on a site. For this “worst-case” estimate of internal dose, it is assumed that 

on the first day of the first year of employment, the covered employee had an acute inhalation intake 

of each of the radionuclides in the source term, in the amounts listed below. 

Based on historical data, it is believed to be highly unlikely that such an intake could have occurred 

without being detected by workplace monitoring at the time. It is also believed that this is a significant 

overestimate of internal dose for an unmonitored covered employee or a covered employee with no 

internal monitoring above detection thresholds. 

3.1.1

Additional assumptions to develop this method are: 

Additional Assumptions

• All intakes are inhalations of standard 5 micrometer AMAD, except for I-131, which is assumed 

to be in vapor form (class SR-1).

• The most soluble form of the radionuclide specified in ICRP 1994a was used to maximize 

dose to systemic organs, except as noted below; dose to lung is not germane to this exercise.

• Because maximum permissible body burdens were the metric (actually uptake) for so many 

years, the assumed implausible uptake was based on a percent of the radionuclide-specific 

Maximum Permissible Body Burden (MPBB) for soluble chemical forms as defined by the 

National Committee on Radiation Protection (NCRP1959). It was assumed that an intake 

resulting in 10% of a MPBB would not have likely occurred to an unmonitored worker or would 

have likely resulted in a readily noticeable bioassay result in a monitored worker, readily 

noticeable air sample, or other indicators of personnel contamination. In other words, an event 

providing the possibility of an intake resulting in a body burden exceeding 10% of the MPBB 

would not have gone unnoticed and there would be some sort of indication in the worker’s 

records. This assumption applies to bona fide DOE sites and national labs with active 

radiation protection programs, not to Atomic Weapons Employers (AWEs). The current ICRP 

methodology is used calculate doses from these implausible intakes.

• For type F and M materials the associated derived intake (i.e., intake resulting in a 10% 

MPBB) was assumed to be 10 and 20 times the 10% MPBB, respectively. The factors of 10 

and 20 come from the current ICRP models and the differences between an intake and the 

activity that is present in the body after the initial clearance of the short-term compartments. 

These factors are used to relate the historical quantity of control, body burden, which was 
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 4 of 15

based on ICRP 2 methods to the present quantity of control, intake, which is based on current 

ICRP methods. These factors were estimated from tables in the November 2002 issue of 

Health Physics that give the intake retention fraction for the whole body (without the ET region) 

as a function of time after acute intake for different elements and inhalation types. Because 

initial deposition in the lung was usually not considered by the ICRP (1959) to be part of the 

“body burden,” the retention fractions used allowed some time for the rapid clearance 

components. Examples are listed in Table 3.1.1-1. 

Table 3.1.1-1. Fractional Retention in Whole Body after Initial Clearance from Lung.a

Element

Fractional Retention in Whole Body After Initial Clearance from Lung

Inhalation Type F Inhalation Type M

3 Days Post Intake 4 Days Post Intake 60 Days Post 

Intake

90 Days Post 

Intake

Strontium 0.227 0.119

Cesium 0.449 0.440

Cerium 0.0683 0.0654

Ruthenium 0.251 0.227 0.0392 0.0392

Barium 0.147 0.104

Lanthanum 0.307 0.291 0.0683 0.0654

Zirconium 0.278 0.252 0.0496 0.0448

Niobium 0.0476 0.0409

Cobalt 0.0410 0.0337

Technetium 0.217 0.160 0.0321 0.0238

Europium 0.0605 0.0568 a

. Source: Potter 2002

• The assumption of type S for Co-58 and Co-60 is used because it results in larger doses to the 

systemic organs because of the high-energy photons. Although the logic in the bullet above 

does not directly apply to type S material, the fractional retention in the whole body is similar 

for type M and type S at 60 and 90 days, so the derived intake is estimated as 20 times the 

10% MPBB.

• Zn-65 is classified as type S in ICRP 68. As for Co-58 and Co-60 type S materials, the logic of 

determining a body burden to intake conversion factor does not directly apply, but use of the 

factor of 20 does not seem unreasonable. 

• Mn-54 type M has a larger dose conversion factor for most organs/tissues and was generally 

more claimant favorable than type F.

• Intakes of tritium are not considered by this method and must be addressed by site, depending 

on the site-specific detection level. Tritium clears very rapidly from the body so multiple 

undetected intakes are possible, although H-3 is readily detected at very small intake levels by 

urinalysis. Tritium dose was often reported with an individual’s external dose prior to 1989; in 

such cases no further H-3 analysis may be necessary.

• To be generic to most of the DOE sites, intakes were assumed to involve the most plausible 

radionuclides for all the sites, even though it is implausible that any one worker had intakes of 

all the radionuclides. Two groups of radionuclides were considered: sites having one or more 

reactors and sites without a reactor. The Nevada Test Site was considered a special case for 

which these two groups of radionuclides might not be appropriate. A review indicated that 

after 1971, the source term at NTS, except in the tunnels, was consistent with that chosen for 

reactor sites. 
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 5 of 15

• For sites without a reactor, the actual ratio of radionuclides in mixtures is not relevant because 

each radionuclide is used at its maximum amount. For sites with reactors, the fission products 

were chosen for the list based on their relative abundance in fuel or contamination after 180 

days of post-irradiation cooling time. This is very conservative for any contamination at 

facilities other than the reactors themselves. 

• A review of the assumed intakes from the method described above and associated possible 

air concentrations for different periods indicate that the uranium intakes based on soluble 

uranium may not have been sufficiently conservative for uranium facilities that controlled their 

programs based on consideration of uranium exposure alone. The derived intakes for 

uranium-234 and uranium-238 were increased by a factor of 100 to ensure that this scenario 

will bound doses to those employees who worked at uranium facilities, but were not included 

in a bioassay program or had no detectable occupationally related activity in them or their 

samples. Derived intake values for U-234 and U-238 respectively may be lowered to 50 nCi 

and 5 nCi for employees who worked in areas that were controlled based on the possible 

presence of other alpha emitters, such as plutonium-238 or plutonium-239. 

Table 3.1.1-2. Radionuclides of interest and respective maximum permissible body burdens, 

assumed intake absorption type, derived maximum plausible intake quantities, and absorption 

type, radiation type and IMBA type and version used for estimating organ doses at DOE sites.

Radionuclide

Soluble

(S)

MPBB 

(nCi) a

Assumed

Intake

Absorption Type

Derived 

Intakeb

Used in 

IMBAc

Runs 

(nCi)

Absorption Type 

Used in 

IMBAc

Runs

Radiation Type IMBAc

Type

IMBAc

Version

Sites without a Reactor:

Sr-90 2,000 F 2,000 F electron > 15 keV OCAS 3.0.48

Tc-99 10,000 F 10,000 F electron > 15 keV OCAS 3.0.63

Cs-137 30,000 F 30,000 F electron > 15 keV NIOSH 1.0.42

U-234 50 F 5,000d F alpha NIOSH 1.0.42

U-238 5 F 500d F alpha NIOSH 1.0.42

Am-241e 50 M 100 M alpha NIOSH 1.0.42

Cm-244 100 M 200 M alpha OCAS 3.0.48

Np-237 60 M 120 M alpha OCAS 3.0.48

Pu-238 40 M 80 M alpha NIOSH 1.0.42

Pu-239 40 M 80 M alpha NIOSH 1.0.42

Th-230f 50 M 100 M alpha OCAS 3.0.63

Cf-252 40 M 80 M alpha OCAS 3.0.63

Sites with a Reactor - All those above plus:

Mn-54 20,000 M 40,000 M Photon>250 keV OCAS 3.0.63

Co-58 30,000 M 60,000 S Photon>250 keV OCAS 3.0.63

Co-60 10,000 M 20,000 S Photon>250 keV NIOSH 1.0.42

Fe-59 20,000 F 20,000 F electron > 15 keV OCAS 3.0.63

Zn-65 60,000 M 120,000 S Photon>250 keV OCAS 3.0.63

Y-91 5,000 M 10,000 M electron > 15 keV OCAS 3.0.63

Nb-95 40,000 M 80,000 M electron > 15 keV OCAS 3.0.59

Zr-95 20,000 F 20,000 F electron > 15 keV OCAS 3.0.59

Ru-103 20,000 F 20,000 F electron > 15 keV OCAS 3.0.59

Ru-106 3,000 F 3,000 F electron > 15 keV OCAS 3.0.59
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 6 of 15

Radionuclide

Soluble

(S)

MPBB 

(nCi) a

Assumed

Intake

Absorption Type

Derived 

Intakeb

Used in 

IMBAc

Runs 

(nCi)

Absorption Type 

Used in 

IMBAc

Runs

Radiation Type IMBAc

Type

IMBAc

Version

I-131 700 F 700 F electron > 15 keV OCAS 3.0.63

Ce-141 30,000 M 60,000 M electron > 15 keV OCAS 3.0.63

Ce-144 5,000 M 10,000 M electron > 15 keV OCAS 3.0.63

Pm-147 60,000 M 120,000 M electron > 15 keV OCAS 3.0.63

Eu-154 5,000 M 10,000 M electron > 15 keV OCAS 3.0.59

Eu-155 70,000 M 140,000 M electron > 15 keV OCAS 3.0.59 a. Source: NCRP (1959). b. Derived maximum plausible intake based on 10% of the soluble NCRP (1959) MPBB multiplied by factors of 10 and 

20 for assumed intake absorption types of F and M, respectively. c. Source: Birchall 2002, 2003, 2003a, and 2003b. d. The uranium values for the general approach are increased to assure that this is a bounding approach for all facilitytypes. The lower values of 50 nCi for U-234 and 5 nCi for U-238 can be used to assign “worst-case” doses to 

individuals who did not work in areas where controls were based primarily on uranium exposure concerns. e.. Pu-241 is accounted for by the Am-241 intake value. f.. Th-230 can be used as a surrogate for Th-232.

After the intakes of the radionuclides were determined, the alpha radionuclides and the beta/gamma 

radionuclides were each summed to show the total intake of each for the reactor and non-reactor 

sites. These alpha and beta/gamma intakes were then used to consider what the average air 

concentrations would have been if the intakes had occurred chronically rather than acutely. A 

breathing rate of 2.4 E+09 milliliters per year, based on Reference Man, was used to calculate the air 

concentrations for 1, 10 and 30 years that would produce the given intakes. Site air concentration 

criteria (area controls, respirator usage, etc.) were based on the permissible air concentration of 

radionuclides assumed to be present in significant quantities. At many facilities, air concentration

controls were specified for alpha, based on Pu-238/239/240, and beta/gamma, based on Sr-90. The 

typical air concentration controls listed in Table 3.1.1-3 are based on the radionuclide air 

concentration values cited in NCRP (1959), AEC (1968), and 10 CFR 835, which for most 

radionuclides did not change significantly over time. In later years, the beta/gamma “limit” may have 

increased to 2 E-9 uCi/ml, but the control criteria also changed and were based on 10% of the limit, 

which would equate to an air concentration of 2 E-10 uCi/ml. 

It can be seen from the Table 3.1.1-3 that even chronic intakes at these levels would exceed typical 

air concentration controls that were being used at most sites since the 1950s to manage access to 

and work practices in areas that had increased radioactivity in air. Acute intakes are assumed for this 

worst-case approach, because these will result in larger assignments of dose.

Table 3.1.1-3 also shows that workers who spent significant time in areas where uranium air 

concentration controls were used to manage access and work practices do not appear to be good 

candidates for the non-uranium facility approaches. The non-uranium facility approaches can be used 

for reconstruction of doses for claims that did not involve exposure in uranium process environments 

for more than a few years. The uranium intakes are included with the alpha intakes with the alpha 

intakes, because it is unlikely that the measured alpha air concentration would have excluded 

uranium. Uranium is shown by itself for nonreactor facilities, because the air concentration control in 

a uranium facility was likely to be larger than the cited gross alpha air concentration control. For 

facilities with reactor source terms the uranium air concentration was not listed, because it is believed 
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 7 of 15

that the control would have been on the gross alpha air concentration and because the uranium 

concentration will be the same as the comparable nonreactor facility.

Table 3.1.1-3. Total Intakes and Possible Average Air Concentrations

Facility 

Type

Radiation 

Type

Intake 

(uCi)

Air 

concentration 

based on 1 

year intake 

(uCi/ml)

Air 

concentration 

based on 10 

year intake 

(uCi/ml)

Air 

concentration 

based on 30 

year intake 

(uCi/ml)

Typical air 

concentration 

controls

(uCi/ml)

NonReactor 

and Nonuranium

Alpha 0.815 3.4E-10 3.4E-11 1.1E-11 2 E-12

Beta/gamma 42 1.8E-08 1.8E-09 5.8E-10 3 E-10

Uranium 0.055 2.3E-11 2.3E-12 7.6E-13 7 E-11

Reactor 

and Nonuranium

Alpha 0.815 3.4E-10 3.4E-11 1.1E-11 2 E-12

Beta/gamma 776 3.2E-07 3.2E-08 1.1E-08 3 E-10

NonReactor 

and 

Uranium

Alpha 6.26 2.6E-09 2.6E-10 8.7E-11 2 E-12

Beta/gamma 42 1.8E-08 1.8E-09 5.8E-10 3 E-10

Uranium 5.5 2.3E-09 2.3E-10 7.6E-11 7 E-11

Reactor 

and 

Uranium

Alpha 6.26 2.6E-09 2.6E-10 8.7E-11 2 E-12

Beta/gamma 776 3.2E-07 3.2E-08 1.1E-08 3 E-10

To encompass employees who may have worked in facilities where uranium was the primary 

radionuclide of concern for internal dose, the assumed uranium intakes were each increased by a 

factor of 100. The factor of 100 was chosen to assure that this approach adequately bounds the level 

of exposure that potentially could have gone unnoticed at uranium facilities.

To facilitate use of these data, two spreadsheets were created based on annual organ doses 

calculated with IMBA (Birchall 2002, 2003, 2003a, and 2003b) for the derived acute inhalation intakes 

shown in Table 3.1.1-2 and for the same derived intakes, except that the U-234 and U-238 intakes are 

lowered respectively to 50 and 5 nanocuries for non-uranium facilities. These spreadsheets allow the 

user to input claim specific data to obtain data necessary for the IREP program for the organs shown 

in Table 3.1.1-4. This approach may be also used for diagnosed conditions that are to have internal 

doses calculated using largest dose to a non-modeled organ as specified in OCAS-TIB-008.

Table 3.1.1-4. Organs Appropriate for Applying the Maximum Internal Dose for 

Certain DOE Complex Claims

Adrenals Esophagus Pancreas Thymus

Brain G. Bladder Skin U. Bladder

Breast Heart Spleen Uterus

Colon Muscle Stomach

4.0

The calculation of organ dose assumes that the covered employee had maximum intakes of all of the 

radionuclides listed in Table 3.1.1-1. While individual exposures to a single radionuclide from the 

Calculation of Internal Dose from Maximum Intakes
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 8 of 15

mixtures defined in this method possibly could be underestimated, despite the attempt to overestimate 

each radionuclide intake, it is believed that the individual dose determined from this method will be 

overestimated, because at a minimum, intakes and doses from 12 radionuclides are considered.

Annual organ doses were calculated as follows. For each radionuclide listed in Table 3.1.1-1, the 

annual organ doses from inhalation intakes were calculated with IMBA (Birchall 2002, 2003, 2003a, 

2003b). The absorption types for Table 3.1.1-1 radionuclides were selected based on the type that 

produced the largest non-metabolic annual organ doses. The ICRP 66 (ICRP 1994) respiratory tract 

model was used with the breathing rate of a standard worker. The assumed particle size distribution 

was 5 μm AMAD (default for worker exposures). The input summary is listed in Table 3.1.1-1. 

To facilitate entry of organ doses into the IREP computer code, Excel® workbooks (entitled 

"Maximum Internal Dose Calculation Workbook.xls” and “Maximum Internal Dose Calculation NonUranium Facility Workbook.xls”) were developed to create the IREP annual organ dose input data. 

Annual organ doses are summed over all radiation types and presented for each year that the 

covered employee was possibly exposed at the DOE site. Examples of annual organ dose output for 

the bladder are shown in Attachment A, Figures A.2.1-1 and A.2.1-2, for sites with and without 

reactors, respectively. In addition, Attachment A contains details of the dose calculations performed 

by the dose estimation Excel® spreadsheet. 

5.0

The following conditions must be met to apply this approach.

Applications and Limitations

• The covered employee’s initial hire date was 1970 or later. 

• The covered employee was not included in a bioassay program or no detectable 

occupationally related activity was found in them or their samples. 

• The target organ must be listed in Table 3.1.1-4 or must be an organ whose dose is based on 

the highest non-modeled organ dose. 

• Employees who would have had no significant exposure to uranium can be assessed with the 

smaller uranium source term. 

• The employment must have been at a DOE site or national laboratory with an active radiation 

protection program.

• Dose from Atomic Weapons Employers (AWEs) must be reconstructed by other methods.

• Dose from Nevada Test Site may be assigned by this approach if the employee worked at 

NTS after 1971, was not involved in tunnel work and meets the other conditions herein. 

• Dose from tritium is to be considered separately. 
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 9 of 15

6.0

10 CFR 835, Occupational Radiation Protection; Final Rule, December 14, 1993. 

References

42 CFR 82, 2002, Methods for Radiation Dose Reconstruction under the Energy Employees 

Occupational Illness Compensation Act of 2000, May.

AEC (Atomic Energy Commission) Manual, Chapter 0524, Standards for Radiation Protection, AEC 

0524-01, November 8, 1968.

Birchall, A, et al, 2003, IMBA ExpertTM-OCAS Edition software, Version 3.0.48, April.

Birchall, A, James AC, IMBA-NIOSH (Phase 1) software, Version 1.0.42, 2002.

Birchall, A, James AC, 2003a, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.59.

Birchall, A, James AC, 2003b, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.63.

EEOICPA (Energy Employees Occupational Illness Compensation Act), 2000, Public Law 106-

398,114 Stat. 1654, 1654A-1231, enacted as Title XXXVI of the Floyd D. Spence National 

Defense Authorization Act for Fiscal Year 2001, October.

ICRP (International Commission on Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, ICRP Publication 2, Pergamon Press, London, 

England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, ICRP Publication 66, Annals of the ICRP Vol. 24 (1-3), Pergamon 

Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994a, Dose Coefficients for Intakes of 

Radionuclides by Workers, Replacement of ICRP Publication 61, ICRP 68, Annals of the ICRP 

Vol. 24 (4), Pergamon Press, Oxford, England.

OCAS-IG-002, Rev.0, “Internal Dose Reconstruction Implementation Guideline,” August 2002.

OCAS-TIB-008, Rev. 0, “Use of ICRP 66 to Calculate Respiratory Tract Doses”, Effective Date: 

9/29/03.

Potter, C. A, 2002, “Intake Retention Fractions Developed from Models Used in the Determination of 

Dose Coefficients Developed for ICRP Publication 68 – Particulate Inhalation,” Health Physics, 

Vol. 83, No. 5, November. 

Recommendation of the National Committee on Radiation Protection, National Bureau of Standards 

(NBS) Handbook 69 (1959), Maximum Permissible Body Burdens and Maximum Permissible 

Concentration of Radionuclides in Air or Water for Occupational Exposure, Superintendent of 

Documents, U.S. Department of Commerce, U.S. Government Printing Office, Washington, 

D.C., June 5, 1959.
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 10 of 15

ATTACHMENT A

A.1 PURPOSE

The purpose of this attachment is to provide a description of the Excel® workbooks entitled 

"Maximum Internal Dose Calculation Workbook.xls" and “Maximum Internal Dose Calculation NonUranium Facility Workbook.xls” that were developed to calculate annual organ doses in support of this 

Technical Information Bulletin (TIB). In addition, this attachment also provides documentation and a 

description of the IMBA (Birchall 2002, 2003, 2003a, 2003b) input and output files used to calculate 

and report organ doses and a description of the user workbook interface developed to assist the dose 

reconstructors in processing claims. 

The workbooks and the IMBA input and output files used in support of the Maximum Internal Dose 

Estimates for Certain DOE Complex Claims TIB are supplied as a folder, “ORAUT-OTIB-0002 Rev 00 

Attachment A,” on CD only. 

A.2 WORKBOOK DESCRIPTION

The Excel® workbooks entitled "Maximum Internal Dose Calculation Workbook.xls" and “Maximum 

Internal Dose Calculation Non-Uranium Facility Workbook.xls” consist of multiple worksheets that are 

grouped according to function. These worksheets and their respective functions are described below.

A.2.1

When a dose reconstructor uses the Maximum Internal Dose Calculation Workbook or Maximum 

Internal Dose Calculation Non-Uranium Facility Workbook, they will only see the worksheet entitled 

"UserPrompt."; all other worksheets contained within the workbook are hidden to assure the dose 

reconstructor enters the correct information in the appropriate location. This worksheet prompts the 

dose reconstructor for the following information:

Dose Reconstructor Interface

• First year of employment

• Organ of interest

• Reactor or non-reactor site

Based on the responses of the dose reconstructor, when the "Open Template" button is pushed, the 

workbook will then calculate the annual organ doses and fill the data template that can be copied into 

an IREP template for the organ of interest. Figures A.2.1-1 and A.2.1-2 below show a typical output 

for a claimant whose first year of employment was 1975. The date of diagnosis of the cancer was 

1985, so the values after this year were manually deleted. The information provided in these 

templates is suitable for direct input into the IREP computer code (NIOSH-IREP 2003) used for 

estimation of probability of causation. Although only 10 years of organ dose information is shown in 

Figures A.2.1-1 and A.2.1-2, the actual template contains 65 years of organ doses to accommodate 

differing years of employment and cancer diagnosis dates.
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 11 of 15

Figure A.2.1-1. Example output file of bladder dose for a covered employee who is assumed to have 

internal exposure beginning in 1975 at a non-uranium DOE complex site with a reactor and whose 

covered condition was diagnosed in 1985.

Enter first exposure year in 

cell H3, for IREP input format

Copy internal dose data from the first year 

1975 through the year of cancer diagnosis

Exposure 

Year

Exposure 

Rate Radiation Type

Dose 

Distribution 

Type

Parameter 

1

Parameter 

2

Parameter 3

1975 chronic alpha Constant 9.85E-02 0 0

1975 chronic electrons E>15keV Constant 1.16E+00 0 0 

1975 chronic photons E>250keV Constant 7.21E-01 0 0

1976 chronic alpha Constant 8.94E-02 0 0

1976 chronic electrons E>15keV Constant 1.28E-01 0 0

1976 chronic photons E>250keV Constant 1.20E-01 0 0 

1977 chronic alpha Constant 8.40E-02 0 0

1977 chronic electrons E>15keV Constant 4.20E-02 0 0

1977 chronic photons E>250keV Constant 2.93E-02 0 0

1978 chronic alpha Constant 7.97E-02 0 0 

1978 chronic electrons E>15keV Constant 2.50E-02 0 0

1978 chronic photons E>250keV Constant 9.48E-03 0 0

1979 chronic alpha Constant 7.69E-02 0 0

1979 chronic electrons E>15keV Constant 1.85E-02 0 0 

1979 chronic photons E>250keV Constant 4.36E-03 0 0

1980 chronic alpha Constant 7.48E-02 0 0

1980 chronic electrons E>15keV Constant 1.47E-02 0 0

1980 chronic photons E>250keV Constant 2.55E-03 0 0 

1981 chronic alpha Constant 7.36E-02 0 0

1981 chronic electrons E>15keV Constant 1.21E-02 0 0

1981 chronic photons E>250keV Constant 1.66E-03 0 0

1982 chronic alpha Constant 7.23E-02 0 0 

1982 chronic electrons E>15keV Constant 1.01E-02 0 0

1982 chronic photons E>250keV Constant 1.13E-03 0 0

1983 chronic alpha Constant 7.14E-02 0 0

1983 chronic electrons E>15keV Constant 8.46E-03 0 0 

1983 chronic photons E>250keV Constant 7.94E-04 0 0

1984 chronic alpha Constant 7.09E-02 0 0

1984 chronic electrons E>15keV Constant 7.13E-03 0 0

1984 chronic photons E>250keV Constant 5.67E-04 0 0 

1985 chronic alpha Constant 7.06E-02 0 0

1985 chronic electrons E>15keV Constant 6.05E-03 0 0

1985 chronic photons E>250keV Constant 4.12E-04 0 0
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 12 of 15

Figure A.2.1-2. Example output file of bladder dose for a covered employee who is assumed to have 

internal exposure beginning in 1975 at a non-uranium DOE complex site without a reactor and whose 

covered condition was diagnosed in 1985.

Enter first exposure year in 

cell H3, for IREP input format

Copy internal dose data from the first year 

1975 through the year of cancer diagnosis

Exposure 

Year

Exposure 

Rate Radiation Type

Dose Distribution 

Type Parameter 1 Parameter 2 Parameter 3

1975 chronic alpha Constant 9.85E-02 0 0

1975 chronic electrons E>15keV Constant 7.14E-01 0 0

1975 chronic photons E>250keV Constant 0.00E+00 0 0

1976 chronic alpha Constant 8.94E-02 0 0

1976 chronic electrons E>15keV Constant 6.84E-02 0 0

1976 chronic photons E>250keV Constant 0.00E+00 0 0

1977 chronic alpha Constant 8.40E-02 0 0

1977 chronic electrons E>15keV Constant 7.08E-03 0 0

1977 chronic photons E>250keV Constant 0.00E+00 0 0

1978 chronic alpha Constant 7.97E-02 0 0

1978 chronic electrons E>15keV Constant 1.00E-03 0 0

1978 chronic photons E>250keV Constant 0.00E+00 0 0

1979 chronic alpha Constant 7.69E-02 0 0

1979 chronic electrons E>15keV Constant 3.59E-04 0 0

1979 chronic photons E>250keV Constant 0.00E+00 0 0

1980 chronic alpha Constant 7.48E-02 0 0

1980 chronic electrons E>15keV Constant 2.51E-04 0 0

1980 chronic photons E>250keV Constant 0.00E+00 0 0

1981 chronic alpha Constant 7.36E-02 0 0

1981 chronic electrons E>15keV Constant 2.05E-04 0 0

1981 chronic photons E>250keV Constant 0.00E+00 0 0

1982 chronic alpha Constant 7.23E-02 0 0

1982 chronic electrons E>15keV Constant 1.72E-04 0 0

1982 chronic photons E>250keV Constant 0.00E+00 0 0

1983 chronic alpha Constant 7.14E-02 0 0

1983 chronic electrons E>15keV Constant 1.45E-04 0 0

1983 chronic photons E>250keV Constant 0.00E+00 0 0

1984 chronic alpha Constant 7.09E-02 0 0

1984 chronic electrons E>15keV Constant 1.23E-04 0 0

1984 chronic photons E>250keV Constant 0.00E+00 0 0

1985 chronic alpha Constant 7.06E-02 0 0

1985 chronic electrons E>15keV Constant 1.05E-04 0 0

1985 chronic photons E>250keV Constant 0.00E+00 0 0
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 13 of 15

A.2.2

The worksheet entitled MPBBs (Maximum Permissible Body Burdens) contains information related to 

the estimation of the derived intakes used for calculation of maximum organ doses. The assumptions 

used for selection of solubility types and intake factors used to derive assumed intakes are provided in 

the main body of the TIB. The worksheet also specifies the radionuclides appropriate for 

consideration at both reactor and non-reactor sites. This worksheet is provided for information only 

and to document the derived intake calculations. The values contained within this worksheet are not 

linked to other worksheets in the workbook.

Maximum Permissible Body Burden

A.2.3

The worksheet entitled "Intakes" lists the radionuclides (Column A, Rows 11 through 38), the IMBA 

absorption and radiation types (Columns B and C, Rows 11 through 38), and derived intake quantities 

for reactor and non-reactor sites (Columns D and E, Rows 11 through 38, respectively) used to 

calculate the maximum annual organ doses. The information related to the radiation-types, 65-year 

doses and radionuclide-specific dose fractions is provided for information only and to assist in the 

quality control process. Thus, this information-only data is not linked to other worksheets nor used 

directly by the worksheet for the organ dose calculations.

Radionuclide Intakes

The annual dose calculations are based on the intake values located in Column F, Rows 11 through 

38. These values are linked to the quantities in Columns D and E, Rows 11 through 38 for reactor 

and non-reactor sites, respectively by conditional statements based on user input from the 

UserPrompt worksheet, which is reproduced on the Intake worksheet at cell B2. The "Assumed 

Intake" values (derived intake quantities) in Column F, Rows 11 through 38 are then linked to the 

respective radionuclide sheets (at location E3) to provide the appropriate multiplier for the organ dose 

calculations (see discussion below).

A.2.4

Individual worksheets are included in the workbook for the 28 radionuclides determined to be 

important to internal dose (see Table 3.1.1-2 of the TIB). Each of these radionuclide-specific 

worksheets contains the derived intake values (cell E3), assumed intake adsorption type used for the 

IMBA runs (cell E2), radiation type (cell A1) and the IMBA type and version used for estimating organ 

doses (cell F1).

Radionuclides Important to Internal Dose

On each of these radionuclide worksheets, results of the IMBA runs (i.e., annual organ dose in rem), 

for a one-nanocurie (nCi) acute inhalation intake of the respective radionuclide, are captured in cells 

B77 through AJ141 for each of the organs of interest (see below). Each of these values is multiplied 

by the derived intake quantity (cell E3) and the products are recorded in cells B6 through AJ70. 

These values represent the annual organ dose resulting from the derived intake values shown in 

Table 3.1.1-2 of the TIB and are summed, by radiation type in the "organ" worksheets that are used 

for the IREP inputs. The IMBA input and output for all radionuclides and organs of interest are 

reproduced on the Attachment A CD.

A.2.5

Because the IREP computer code requires specification of radiation type, three worksheets were 

developed to add the IMBA doses from the radionuclide worksheets for each radiation type. The 

radiation type worksheets are entitled "Alpha,'" "Electron>15 keV," and "Photon>250keV." 

Radionuclides contributing dose to each of these radiation type worksheets are as follows:

Radiation Types
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 14 of 15

• Alpha - Cf-252, Am-241, Cm-244, Np-237, Pu-238, Pu-239, U-234, and U-238

• Electron>15keV - Sr-90, Cs-137, I-131, Ce-141, Eu-154, Eu-155, Nb-95, Ru-103, Ru-106, Zr 

95, Y-91, Tc-99, and Pm-147

• Photon>250keV - Co-58, Co-60, Mn-54, and Zn-65

For each of the three radiation types, annual doses from the appropriate radionuclide worksheets 

(cells B6 through AJ70) were summed in cells B6 through AJ70 for each organ of interest. These 

annual doses summed by radiation type are then carried to the organ worksheets for preparation of 

the IREP input template.

A.2.6

Individual worksheets were included for each of the following organs of interest:

Organs of Interest

• Adrenals

• Esophagus

• Pancreas

• Thymus

• Brain

• G. Bladder

• Skin

• Thyroid

• Breast

• Heart

• Spleen

• U. Bladder

• Colon

• Muscle 

• Stomach

• Uterus

These organ worksheets arrange the organ doses by radiation type from all of the radionuclides of 

interest into a format suitable for direct entry into the IREP computer code (see Figures A.2.1-1 and 

A.2.1-2 above). The annual dose information by radiation type is linked from the radiation type 

worksheets and presented in columns B, C, and D; Rows 6 through 70, representing 65 years of 

exposure. These doses are then arranged in the IREP input form (shown in various colors) with three 

entries (one for each radiation type) for each year of exposure. These radiation- and year-specific 

annual doses (rem) are shown in Column J, Rows 6 through 200. The first year that dose is 

calculated is shown in cell H3 and is set based on input provided by the dose reconstructor through 

the UserPrompt worksheet. 

To setup the IREP input data, the dose reconstructor would copy cells F6 through L200 and paste 

these values into the identical cells in the IREP template. The actual number of yearly doses 

considered in the probability of causation calculation is set by the reconstructor in the "Number of 

Exposures" block in the IREP template.
Effective Date: 11/07/2003 Revision No. 00 Procedure No. ORAUT-OTIB-0002 Page 15 of 15

A.3 REFERENCES

Birchall, A, et al, 2003, IMBA ExpertTM-OCAS Edition software, Version 3.0.48, April.

Birchall, A, James AC, IMBA-NIOSH (Phase 1) software, Version 1.0.42, 2002.

Birchall, A, James AC, 2003a, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.59.

Birchall, A, James AC, 2003b, IMBA ExpertTM-OCAS Edition software, Part 3 - Split SEEs, Version 

3.0.63.

NIOSH-IREP, 2003, Interactive RadioEpidemiological Program version 5.2.1, National Institute of 

Occupational Safety and Health, Cincinnati, OH, May